vice president - operations entergy nuclear operations ... · vice president - operations entergy...

July 25, 2002

Mr. Fred DacimoVice President - OperationsEntergy Nuclear Operations, Inc.Indian Point Nuclear Generating Units 1 & 2295 Broadway, Suite 1Post Office Box 249Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT 2 - NRC INSPECTION REPORT 50-247/02-04

Dear Mr. Dacimo:

On June 29, 2002, the NRC completed an inspection at the Indian Point 2 Nuclear Power Plant. The enclosed report presents the results of that inspection. The results were discussed on July 3, 2002, with members of your staff.

The inspection was an examination of activities conducted under your license as they relate tosafety and compliance with the Commission’s rules and regulations, and with the conditions ofyour license. The inspection also reviewed the security program and recent emergency plandocument changes. Within these areas, the inspection consisted of a selected examination ofprocedures and representative records, observations of activities, and interviews withpersonnel. Based on the results of this inspection, the inspectors identified two issues of verylow safety significance (Green).

The NRC has increased security requirements at Indian Point 2 Nuclear Power Plant inresponse to terrorist acts on September 11, 2001. Although the NRC is not aware of anyspecific threat against nuclear facilities, the NRC issued an Order and several threat advisoriesto commercial power reactors to strengthen licensees’ capabilities and readiness to respond toa potential attack. The NRC continues to monitor overall security controls and will issuetemporary instructions in the near future to verify by inspection the licensee's compliance withthe Order and current security regulations.

Mr. Fred Dacimo 2

In accordance with 10 CFR 2.790 of the NRC’s "Rules of Practice," a copy of this letter and itsenclosure will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records (PARS) component of the NRC’s documentsystem (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Should youhave any questions regarding this report, please contact Mr. Peter Eselgroth at 610-337-5234.

Sincerely,

Brian E. Holian, Deputy DirectorDivision of Reactor Projects

Docket No.50-247License No. DPR-26

Enclosure: Inspection Report 50-247/02-04

Attachment 1 - Supplemental Information

cc w/encl: J. Yelverton, Chief Executive OfficerM. R. Kansler, Senior Vice President and Chief Operating OfficerJ. Herron, Senior Vice PresidentR. J. Barrett, Vice President - OperationsL. Temple, General Manager - OperationsD Pace, Vice President - EngineeringJ. Knubel, Vice President Operations SupportJ. McCann, Manager, Nuclear Safety and Licensing J. Kelly, Director of LicensingC. Faison, Manager - Licensing, Entergy Nuclear Operations, Inc.H. Salmon, Jr., Director of Oversight, Entergy Nuclear Operations, Inc.J. Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc.W. Flynn, President, New York State Energy, Research and Development AuthorityJ. Spath, Program Director, New York State Energy Research and Development AuthorityP. Eddy, Electric Division, New York State Department of Public ServiceC. Donaldson, Esquire, Assistant Attorney General, New York Department of LawT. Walsh, Secretary, NFSC, Entergy Nuclear Operations, Inc.Mayor, Village of BuchananR. Albanese, Executive Chair, Four County Nuclear Safety CommitteeS. Lousteau, Treasury Department, Entergy Services, Inc.M. Slobodien, Director Emergency ProgramsB. Brandenburg, Assistant General CounselP. Rubin, Operations Manager

Mr. Fred Dacimo 3

Assemblywoman Sandra Galef, NYS AssemblyCounty Clerk, Westchester County LegislatureA. Spano, Westchester County ExecutiveR. Bondi, Putnam County ExecutiveC. Vanderhoef, Rockland County ExecutiveE. A. Diana, Orange County ExecutiveT. Judson, Central NY Citizens Awareness NetworkM. Elie, Citizens Awareness NetworkD. Lochbaum, Nuclear Safety Engineer, Union of Concerned ScientistsPublic Citizen’s Critical Mass Energy ProjectM. Mariotte, Nuclear Information & Resources ServiceF. Zalcman, Pace Law School, Energy ProjectL. Puglisi, Supervisor, Town of CortlandtCongresswoman Sue W. KellyCongressman Ben GilmanCongresswoman Nita LoweySenator Hilary Rodham ClintonSenator Charles SchumerJ. Riccio, GreenpeaceA. Matthiessen, Executive Director, Riverkeepers, Inc.M. Kapolwitz, Chairman of County Environment & Health CommitteeA. Reynolds, Environmental AdvocatesM. Jacobs, Executive Director, Westchester Peoples Action CoalitionD. Katz, Executive Director, Citizens Awareness NetworkP. Gunter, Nuclear Information & Resource ServiceP. Leventhal, The Nuclear Control InstituteK. Copeland, Pace Environmental Litigation ClinicR. Witherspoon, The Journal News

Mr. Fred Dacimo 4

Distribution w/encl: H. Miller, RA/J. Wiggins, DRA (1)H. Nieh, RI EDO CoordinatorP. Habighorst, SRI - Indian Point 2S. Richards, NRR (ridsnrrdlpmlpdi)P. Eselgroth, DRPP. Milano, PM, NRRG. Vissing, PM, NRR (Backup)S. Barber, DRPW. Cook, DRPR. Junod, DRPR. Martin, DRPRegion I Docket Room (w/concurrences)

DOCUMENT NAME: C:\ORPCheckout\FileNET\ML022100476.wpdAfter declaring this document “An Official Agency Record” it will be released to the Public. Toreceive a copy of this document, indicate in the box: "C" = Copy withoutattachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy

OFFICE RI/DRP RI/DRP E RI/DRP ENAME PHabighorst/ PEselgroth/ BHolian/DATE 7/08/02 7/ /02 7/ /02

OFFICIAL RECORD COPY

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No. 50-247

License No. DPR-26

Report No. 50-247/02-04

Licensee: Entergy Nuclear Operations, Inc..

Facility: Indian Point 2 Nuclear Power Plant

Location: Buchanan, New York 10511

Dates: May 12 - June 29, 2002

Inspectors: Peter Habighorst, Senior Resident InspectorLois James, Resident InspectorWilliam Raymond, Senior Resident Inspector, Pilgrim Greg Smith, Security SpecialistAnthony Dimitriadis, Security Specialist Todd Fish, Operations EngineerDavid Silk, Senior Emergency Preparedness Inspector

Approved by: Peter W. Eselgroth, ChiefProjects Branch 2Division of Reactor Projects

ii

SUMMARY OF FINDINGS

IR 05000247-02-04, on 5/12 - 6/29/2002, Entergy Nuclear Operations, Inc.; Indian Point 2Nuclear Power Plant. Barrier Integrity and Mitigating Systems.

The report covered a seven week period of inspection by resident and region-based inspectors. No findings of significance were identified. The NRC’s program for overseeing the safeoperation of commercial nuclear power reactors is described at its Reactor Oversight Processwebsite at http://www.nrc.gov/reactors/operating/oversight.html.

Cornerstone: Barrier Integrity

Green. On May 27, 2002, during surveillance testing of the safety injection discharge motor-operated valve (851B), the valve failed to stroke closed. The initial operability evaluation did notconsider the non-automatic containment isolation function for this valve. This event wasdocumented in condition report No. 200205433. The performance issue associated with thisfinding is a weakness in operator knowledge of multi-function safety system components. Thisis the second recent example where operators did not consider this function for a safety-relatedvalve. The first example was documented in NRC report 50-247/2002-003, section 1R15. Theuntimely and incomplete operability assessment for safety injection discharge valve 851B hasvery low safety significance since the containment isolation valve was restored to an operablestatus prior to exceeding Technical Specification 3.6.A.3.a.2.d limiting condition for operation.

Cornerstone: Mitigating Systems

Green. On May 17, 2002, multiple grounds on the protective circuit for Unit 1 substation102NS3 resulted in a loss of the 13.8 kilovolt (kv) lighting and power bus section 3. Theconsequence of this event was a loss of alternate safe shutdown power to all major alternatesafe shutdown pumps and selected instrumentation. At the time, the Unit 2 normal andemergency electrical power supplies were available to supply power to the above statedmitigation equipment and instrumentation. The licensee repaired and restored the 13.8 kv bussection 3 within 30 hours of the fault. The performance issue is inadequate retirement ofprotective circuits for 440 volt substations (132PC3 and 142PC3) that could impact availabilityof alternate safe shutdown power supplies. This issue is more than minor since unavailability ofalternate safe shutdown equipment for 30 hours is viewed as a precursor to a significant eventand the alternate safe shutdown power supplies are a risk-significant maintenance rule systemwhich was unavailable for greater than 24 hours.

iii

TABLE OF CONTENTS

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

Report Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

SUMMARY OF PLANT STATUS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1. REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11R04 Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11R06 Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31R11 Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41R12 Maintenance Rule Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41R13 Maintenance Risk Assessment and Emergent Work Activities . . . . . . . . . . . . . . 51R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71R16 Operator Workarounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81R17 Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91R19 Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101R23 Temporary Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111EP4 Emergency Action Level and Emergency Plan Changes . . . . . . . . . . . . . . . . . 11

3. Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123PP1 Access Authorization Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123PP2 Access Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

4. OTHER ACTIVITIES (OA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 134OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 134OA3 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144OA4 Inspection Item Follow-up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

Key Points of Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

List of Items Opened, Closed, and Discussed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

List of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

Report Details

SUMMARY OF PLANT STATUS

The plant operated at full power during the inspection period.

1. REACTOR SAFETY(Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, EmergencyPreparedness)

1R04 Equipment Alignment

Partial System Walkdowns

a. Inspection Scope (71111.04)

On May 14, 2002, the inspector performed a partial system walkdown of the 21 and 22trains of the auxiliary feedwater system, when the 23 auxiliary feedwater (AFW) pumpwas out of service for surveillance testing and predictive maintenance. The inspectorobserved the conditions in the auxiliary feedwater building to verify no discrepancieswould impact the AFW system function. The inspector also reviewed the status of keyAFW system components based on check off list (COL) 21.3, “Steam Generator WaterLevel and Auxiliary Boiler Feedwater," revision 22. The inspector observed the physicalcondition of the 21 and 22 AFW pumps, reviewed the operations logs, and discussedperformance issues with an operator.

On June 20, 2002, the inspector performed a partial system walkdown of the 21 and 23trains of the charging system. The review was performed while the 22 charging pumpwas isolated and tagged out (reference tagout 16488) for corrective maintenance. Theinspector reviewed the status of equipment alignment based upon COL 3.1, Chemicaland Volume Control System,” revision 33, and plant drawing nos. 9321-F-2736-114,“Flow Diagram Chemical and Volume Control System, sheet 1,” and A208168-51, FlowDiagram Chemical and Volume Control System, sheet 2.”

b. Findings

No significant findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours


The inspector toured the areas important to plant safety and risk, based upon a review ofSection 4.0, “Internal Fires Analysis,” and Table 4.6-2, “Summary of Core DamageFrequency Contributions from Fire Zones,” in the Indian Point 2 Individual PlantExamination for External Events (IPEEE). The inspector evaluated conditions related to(1) licensee control of transient combustibles and ignition sources; (2) the materialcondition, operational status, and operational lineup of fire protection systems, equipment

2

and features; and (3) the fire barriers used to prevent fire damage or fire propagation. The areas reviewed were:

• Fire Zone 74A, Electrical Penetration Area• Fire Zone 39A, Turbine Building 33 foot elevation• Fire Zone 11, Cable Spreading Room

Reference material consulted by the inspector included the Fire ProtectionImplementation Plan, Pre-Fire Plan, and Station Administrative Orders (SAOs)-700, “FireProtection and Prevention Policy,” SAO-701, “Control of Combustibles and Transient FireLoad,” SAO-703, “Fire Protection Impairment Criteria and Surveillance,” and CalculationPGI-00433, “Combustible Loading Calculation.” The inspector reviewed licenseecorrective actions to address fire protection program deficiencies for the above stated firezones.

b. Findings


.2 Fire Drills


On June 20, 2002, the inspector observed an announced fire brigade drill. The drill was in accordance with the pre-planned drill scenario for a 23 emergency diesel generator(EDG) control panel fire. This fire brigade drill was a training drill for a new fire brigadeleader with the shift fire brigade leader acting in an advisory role. The purpose of thisobservation was to evaluate the readiness of the licensee's personnel to prevent andfight fires. The inspector evaluated the following aspects:

• Protective clothing/turnout gear is properly donned. • Self-contained breathing apparatus (SCBA) equipment is properly worn and used.• Fire hose lines are capable of reaching all necessary fire hazard locations, are

laid out without flow constrictions, and are simulated being charged with water.• Fire area is entered in a controlled manner.• Sufficient fire fighting equipment is brought to the scene by the fire brigade.• Effective smoke removal operations are simulated.• The fire fighting pre-plan strategies are utilized.• The licensee’s pre-planned drill scenario is followed.• The drill objectives’ acceptance criteria are met.

Minor deficiencies that did not impact the ability to fight a fire were noted during the drillcritique and were entered into the Condition Reporting System (CR Nos. 200206616 and200206187.

b. Findings


3

1R06 Flood Protection Measures


The inspector reviewed and toured areas containing equipment used to mitigate anddetect an internal flood on various elevations within the primary auxiliary building.

The plant areas selected contained risk significant equipment based on the IndividualPlant Examination of External Events (IPEEE) Section 5.0. Specifically, internal floodinitiations from service water, fire protection line breaks, and refueling water storage tankbreaks in the primary auxiliary building contribute approximately 6% of the overall coredamage frequency from internal floods. The inspection determined whether procedureswere adequate, mitigation systems were operable pursuant to technical specificationrequirements, and sump level instruments were properly tested.

The inspector reviewed applicable licensee procedures and surveillance procedures,which included actions to mitigate the effects of flooding and tours to verify operability ofmitigating equipment. The procedures reviewed are listed in Attachment 1.

The inspector reviewed Updated Final Safety Analysis Report section 9.3.3.2.2. Theinspector also reviewed safety evaluation NS-2-84-047, Primary Auxiliary Building FloodAlarms, associated with a modification to the sump level switches and alarms.

b. Findings

No significant findings were identified. 1R07 Heat Sink Performance

a. Inspection Scope (71111.07A)

The inspector verified that the licensee’s program was adequate to ensure proper heatexchanger performance for the 21 and 22 emergency diesel generator (EDG) jacketwater and lube oil coolers. The references used for this inspection included theEmergency Diesel Generator System Health Report, the EDG Basis Document revision2, and Heat Exchanger Inspection Reports for Work Order Nos. 01-22146, 01-22147, 98-02651 and 98-02653.

The inspector reviewed heat exchanger preventive maintenance (PM) records to verifythat the performance monitoring techniques ensuring heat removal capabilities wereacceptable. The inspector verified that the inspection results were appropriatelycompared against established acceptance criteria; the performance monitoringconsidered the differences between plant conditions and design conditions established inCalculation PGI-00354-02; the frequency of testing and inspections was sufficient; and,the licensee established acceptance criteria for bio-fouling control. The inspector verifiedthat the results were evaluated to ensure proper heat exchanger operation.

The inspector reviewed a sample of corrective action system condition reports related tothe selected equipment to verify that identified problems were appropriately resolved

4

(reference CR Nos. 199808862, 200110318 and 200203601). The inspector alsoconducted a walkdown of the selected heat exchangers to assess material condition.

b. Findings


1R11 Licensed Operator Requalification Program

a. Inspection Scope (71111.11Q)

Background

On May 31, 2002, a staff crew failed the licensee administered final simulator scenario ofthe accelerated high intensity training program. The staff crew performance wasdocumented in CR No. 200205647. The operators were prohibited from fulfilling licensedduties until successfully completing a two week remediation program. The high intensitytraining program was part of the training improvement plan that was documented in NRCreport 50-247/02-09.

a. Inspection Scope (71111.11Q)

On June 21, 2002, the inspector observed the performance of the staff crew duringscenario ESR-022-06. The inspector verified that the scenario met the attributes outlinedin Attachment 11 of Inspection Procedure 71111.11. The training staff administered thescenario and evaluated operator performance. The operators passed the evaluation andthe inspector’s evaluation agreed with that of the facility.

b. Findings

No significant findings were identified. 1R12 Maintenance Rule Implementation


The inspector reviewed risk significant equipment problems that were associated with theauxiliary feedwater, service water, 13.8 kilovolt system, and 440 volt systems. Theinspector reviewed licensee follow-up actions to assess the effectiveness of maintenanceactivities. Issues selected for review included licensee identification of any maintenancepreventable functional failures and repetitive failures, as well as, problem identificationand resolution of any maintenance related issues. The inspector also reviewed systemavailability, system reliability monitoring, and system engineering involvement. Theinspector reviewed the maintenance rule basis documents as listed in Attachment 1. Thefollowing specific issues were reviewed:

• CR No. 200111489, Failure of automatic function of auxiliary feedwater flowcontrol valve FCV-406A.

5

• CR No. 200202589, Failure of valve SWN-7 (service water to turbine lube oilcoolers) to fully close due to operator gear stripping.

• CR No. 200205115, Loss of 13.8 KV Alternate Safe Shutdown Bus Section 3.

The inspector reviewed the licensee’s evaluations for the deficiencies identified inCondition Report 200205115, which concerned the loss of the power supply for alternatesafe shutdown equipment (via substations 12RW3 and 12FD3) for 29 hours and 42minutes (see report detail 1R13). The inspector reviewed the licensee’s correctiveactions to assess the impact of the equipment loss on the reliability and availability of the440 volt system and to assess the event as a maintenance preventable functional failure,since the failure of a “retired” IP Unit 1 protective relaying circuit affected a Unit 2 risksignificant system.

b. Findings


1R13 Maintenance Risk Assessment and Emergent Work Activities


The inspector observed selected portions of emergent maintenance work activities toassess the licensee’s risk management in accordance with 10 CFR 50.65 (a)(4). Theinspector verified that the licensee took the necessary steps to plan and control emergentwork activities, to minimize the probability of initiating events, and to maintain thefunctional capability of mitigating systems. The inspector discussed the riskmanagement with maintenance and operations personnel for the following work orders(WOs):

• WO 02-02772, Safety Injection Valve 851B Failed to Stroke (CR 200205369).• WO 02-02812, Control Room West Wall Inspections (CR 200205807).• WO 02-45258, Unit 1 Lighting and Power Bus Section 3 Overcurrent Relay

Retired Input Bypass (CR 200205115).• WO 02-00735, Valve MS-41 Main Steam Supply to 22 Auxiliary Feed Pump

packing leak.

The inspector reviewed licensee actions to assess, evaluate, and correct the deficiencyin Condition Report 200205369, which concerned the failure of safety injection valveMOV-851B to stroke during a surveillance test. Further NRC review of this matter isdescribed in section 1R15 of this report.

On May 17, the inspector observed operators responding to a loss of the Unit 1 13.8kilovolt lighting and power bus section 3. The inspector observed repair activities and evaluated the licensee’s causal evaluation and corrective actions as documented inCondition Report 200205115.

b. Findings

6

GREEN. On May 17, multiple grounds on the over-current protective circuit for Unit 1substation 102NS3 resulted in a loss of the 13.8 kv lighting and power (L&P) bus section3. The consequence of this event was a loss of available power to major alternate safeshutdown (10 CFR 50 Appendix R section III, Specific Requirements) pumps (23 and 24service water pumps, 23 component cooling water pumps, 21 auxiliary boiler feedwaterpump, 23 charging pump, and 21 and 22 residual heat removal pumps). In addition,some instrumentation for alternate safe shutdown was de-energized, including thealternate safe shutdown source range nuclear instrument and reactor coolant systemloop 21 and 22 hot and cold leg temperature indicators.

This issue is more than minor since unavailability of alternate safe shutdown equipmentfor 30 hours is viewed as a precursor to a significant event. The alternate safe shutdownpower supply is considered a risk-significant maintenance rule system which wasunavailable for approximately 30 hours, greater than guidance contained in NRC ManualChapter 0609 Appendix A (24 hours). The inspector consulted Region 1 senior riskassessment personnel to further characterize the risk significance of this performanceissue. The NRC evaluated all accident sequences in the licensee’s Individual PlantExamination for External Events (IPEEE) Table 4.6, in which alternate safe shutdownequipment is assumed for recovery action. The results of the analysis, based upon aduration of 30 hours, resulted in a very low risk significance (5E-7/reactor-year coredamage frequency). The dominant sequence is a fire in the control room which results inde-energizing the Unit 2 normal and emergency sources of power.

The licensee repaired and restored the 13.8 kv bus section 3 within 30 hours of the initialfault. The licensee implemented appropriate risk assessments and adhered toadministrative requirements identified in station administrative order 703, “Fire ProtectionImpairment Criteria and Surveillance.” The administrative action for alternate safeshutdown components was to restore them to an operable condition within 72 hours. The Unit 2 normal and emergency electrical power supplies were available to supplypower to the above stated mitigation equipment and instrumentation. During the courseof troubleshooting the electrical grounds, the licensee identified that two additional over-current circuits were energized to long-standing retired 440 volt substations (132PC3 and142PC3) at Unit 1. The protective circuits supply input into the 13.8KV supply breaker(SB1-3) to L&P bus section 3. No preventative maintenance is performed on the retiredenergized protective circuits at Unit 1. Restoration of the L&P bus section 3 wasaccomplished with temporary facility change TFC 2002-044 that permanently disabledthe overcurrent protection circuit to substations 132PC3 and 142PC3 and bypassed thegrounded sections of protection from substation 102NS3. The licensee plans an extent-of-condition review to evaluate all 440 volt Unit 1 substations by the end of September,and to identify if protective circuits to retired equipment are still energized.

1R15 Operability Evaluations


The inspectors reviewed selected operability determinations to assess the adequacy ofthe evaluations, the use and control of compensatory measures, compliance with theTechnical Specifications, and the risk significance of the issues. The inspectors used theTechnical Specifications, Technical Requirements Manual, Final Safety Analysis Report,

7

and associated Design Basis Documents as references. The specific issues reviewedincluded:

• CR No. 200203319, Stroke Time for Post Accident Containment Vent Valve E-3.• CR No. 200205369, Loose Wire in MOV 851B Open Circuit.• CR No. 200205673, Large Amount of Debris on the Floor of the 480V Switchgear

Room.

b. Findings

GREEN. On two occasions between May 27 and 29, 2002, during surveillance testing ofsafety injection discharge motor-operated valve (851B) the valve failed to stroke closed. The initial operability evaluation performed by operations did not consider thecontainment isolation function for this valve. Valve 851B has three safety-relatedfunctions. The first valve function is to open and direct flow from the high head safetyinjection pump 22 to the appropriate safety injection header. The second function ofvalve 851B is to close upon a failure of the 21 safety injection pump to ensureequalization of safety injection flow. The third function of valve 851B is to be remotelyclosed by operators for containment isolation following a transfer from low head cold legrecirculation to hot leg recirculation, as documented in emergency operating procedure(EOP) ES1.4, “Transfer to Hot Leg Recirculation.” For 9.25 hours on May 27, and forapproximately 1.5 hours on May 29, operators did not consider the non-automaticcontainment isolation function and did not enter Technical Specification (TS) 3.6.A.3.a. Following inspector discussions with the on-shift senior reactor operator on May 29,actions were taken in accordance with TS 3.6.A.3.a.

The containment isolation function was restored to an operable status by a change toEOP ES 1-4 that provided guidance to operators to close the valve from its motor controlcenter with the appropriate tools and equipment in the primary auxiliary building. Theinspector verified the EOP change and staging of tools and equipment in the primaryauxiliary building. This event was documented in CR No. 200205433.

The performance issue associated with this finding was the lack of operator awarenessof the multiple safety functions for mitigating systems valves. This was the secondrecent example of operators not considering all the safety functions for mitigating systemvalves. The first example was documented in NRC report 50-247/2002-003 section1R15.

The incomplete operability assessment for safety injection discharge valve 851B has verylow safety significance since the containment isolation valve was restored to an operablestatus prior to exceeding the 36-hour allowed outage time, per TS 3.6.A.3.a.2.d. Consequently, no violation of TS 3.6.A.3.a.2.d occurred.

1R16 Operator Workarounds


The inspector reviewed the licensee’s list of twenty operator burdens as of

8

June 25, 2002, to assess the cumulative effects on system reliability, availability, andpotential for mis-operation of a system. The inspector evaluated the below listedburdens to assess the operator impact to implement abnormal operating procedures oremergency operating procedures:

• CR No. 200109397, Nuclear Detector Rate Comparator Indication.• CR No. 200109706, Letdown Pressure Regulator Manual Operation.• CR No. 200110544, Residual Heat Removal Flow Control Valve Position

Indication. • CR No. 200202395, 24 Reactor Coolant Pump Motor Temperature Indication.• CR No. 200101809, 22 Main Feedwater Pump Steam Stop Valve Control Switch.

The inspector verified that none of the deficiencies were an operator work-around thatwould require operator action during a transient that was time-dependent and wouldadversely affect risk or major equipment reliability. The inspector reviewed thecumulative effect of the burdens for either increasing the potential for initiating events oradversely impacting mitigating systems. The inspector verified the condition of thedeficiencies and the compensatory measures taken. The inspector used OperationsAdministrative Steplist (OASL)-15.43, “Operator Burden Program”, Revision 0, as areference for this review.

b. Findings


1R17 Permanent Plant Modifications

a. Inspection Scope (71111.17A)

The inspection consisted of a review of plant modification FMX-97-12648-M, “Installationof Canopy Seal Clamp Assemblies,” and the associated safety evaluation No. 01-0707-M. This modification is being planned for implementation during the upcoming refuelingoutage as a contingency if a reactor vessel head penetration nozzle for a spare controlrod drive mechanism (CRDM) was found to be leaking. Five canopy seal leaks havebeen repaired at Indian Point Unit 2 since the issue was first identified in the mid 1980s. The most recent repair was performed in 1997. The purpose of this modification is toprovide a seal clamp around the leaking canopy seal. This modification has beenimplemented at a number of other nuclear facilities in the recent past.

The specific inspection attributes evaluated included:

• compatibility of clamp assembly with reactor coolant system.• environmental qualification and evaluation of seismic interactions with the

spare CRDM nozzles.• verification that reactor coolant pressure boundary was not compromised.• verification of post-maintenance testing proposed.• verification of failure modes introduced by the modification.

9

This contingency modification has the potential to impact the barrier integritycornerstone.

b. Findings


1R19 Post Maintenance Testing


The inspector reviewed post-work test (PWT) procedures and associated testingactivities to assess whether: 1) the effect of testing in the plant had been adequatelyaddressed by control room personnel; 2) testing was adequate for maintenanceperformed; 3) acceptance criteria were clear and adequately demonstrated operationalreadiness consistent with design and licensing documents; 4) test instrumentation hadcurrent calibrations, range, and accuracy for the application; and 5) test equipment wasremoved following testing.

The selected testing activities involved components that were risk significant as identifiedin the IP2 Individual Plant Examination. The regulatory references for the inspectionincluded Technical Specification 6.8.1.a. and 10 CFR 50, Appendix B, Criteria XIV,“Inspection, Test, and Operating Status.” The following testing activities were evaluated:

• PWT IP2-02-48106 Replacement of Zone 2 Weld Channel ContainmentPressurization system valve 1110-8, performed on June 25, 2002.

• PWT IP2-02-42774 Replace Packing and Change Out Stuffing Boxes for the 22Charging Pump, performed on June 20-24, 2002.

b. Findings


1R22 Surveillance Testing


The inspector reviewed surveillance test procedures and observed testing activities toassess whether: 1) the test preconditioned the component tested; 2) the effect of thetesting was adequately addressed in the control room; 3) the acceptance criteriademonstrated operational readiness consistent with design calculations and licensingdocuments; 4) the test equipment range and accuracy was adequate and the equipmentwas properly calibrated; 5) the test was performed per the procedure; 6) the testequipment was removed following testing; and 7) test discrepancies were appropriatelyevaluated. The surveillances observed were based upon risk significant components asidentified in the Indian Point 2 Individual Plant Examination. The regulatory requirementsthat provided the acceptance criteria for this review were 10 CFR 50, Appendix B,Criterion V, “Instructions, Procedures, and Drawings,” Criterion XIV, “Inspection, Test,

10

and Operating Status,” Criterion XI, “Test Control,” and Technical Specification 6.8.1.a. The following test activities were reviewed:

• PT-Q27B, 23 Auxiliary Feedwater Pump Test, Revision 8 (CR 200204977 and200204979), performed on May 14, 2002.

• PT-Q13 Data Sheets 174, LCV-1208A, and 176, LCV-108B, Inservice Valve Test,Revision 23, performed on May 22, 2002.

• PT-M93, Fuel Storage Building Filtration System Functional Test, Revision 0,performed on June 13, 2002.

b. Findings


1R23 Temporary Plant Modifications


The inspector reviewed temporary facility change (TFC) package, TFC-2001-100, “GasTurbine 1 Black-start Diesel Battery Installation.” This TFC was prepared to install atemporary replacement to the installed black-start emergency diesel generator batterydue to a failure in December 2001. The inspector reviewed the TFC and the associatedsafety evaluations to verify the facility change did not adversely impact safety systemoperability and the license requirements, and did not violate 10 CFR 50.59. Theinspector performed a field walkdown, reviewed the site procedure for performingtemporary facility changes, SAO 206, Revision 22, “Temporary Field Change,” andevaluated the comprehensiveness of the licensee’s review of impacted procedures anddrawings.

On May 20, 2002, the inspectors observed the installation of TFC 2002-033, “FCV-1207Hot Tap Isolation Valve.” TFC 2002-033 installed a hot tap stop valve down stream offlow control valve (FCV)-1207, low pressure steam dump, in order to isolate FCV-1207from the condenser and perform corrective maintenance on FCV-1207. This TFC wasselected for observation due to the potential to degrade main condenser vacuum duringthe hot tap installation. The inspectors reviewed the temporary facility change package,including the 10 CFR 50.59 screening and Work Order #IP2-02-42003 under which theTFC was installed, against the Updated Final Safety Analysis Report and StationAdministrative Order SAO-206, “Temporary Facility Change,” Revision 22.

b. Findings


1EP4 Emergency Action Level and Emergency Plan Changes


The inspector conducted an in-office review of licensee submitted changes for theemergency plan-related documents to determine if the changes decreased the

11

effectiveness of the plan. A thorough review was conducted of documents related to therisk significant planning standards (RSPS), such as classifications, notifications, andprotective action recommendations. These changes (final changes) were reviewedagainst 10 CFR 50.54(q) to ensure that the changes do not decrease the effectiveness ofthe plan, and that the changes made continue to meet the standards of 10 CFR 50.47(b)and the requirements of Appendix E. These changes are subject to inspections toensure that the changes continue to meet NRC regulations. The submitted and revieweddocuments (Plan and Implementing Procedures) are listed in Attachment 1.

b. Findings

No findings of significance were identified.

3. Safeguards(Cornerstone: Physical Protection)

3PP1 Access Authorization Program


The below listed activities were conducted to determine the effectiveness of thelicensee’s behavior observation portion of the personnel screening and fitness-for-duty(FFD) programs as measured against the requirements of 10 CFR 26.22 and thelicensee’s fitness-for-duty program documents.

Five supervisors representing the Indian Point Units 2 and 3, Maintenance, OperationsProcedures, Operations, and Engineering departments were interviewed on May 22,2002, regarding their understanding of behavior observation responsibilities and theability to recognize aberrant behavior traits. Two Access Authorization/Fitness-for-Dutyself-assessments, two semi-annual fitness-for-duty performance data reports, an audit,and event reports and loggable events for the four previous quarters were reviewedduring May 20-24, 2002. On May 22, 2002, five individuals who perform escort dutieswere interviewed to establish their knowledge level of those duties. Behavior observationtraining procedures and records were reviewed on May 21, 2002.

b. Findings


3PP2 Access Control


The below listed activities were conducted during the inspection period to verify that thelicensee has effective site access controls, and equipment in place designed to detectand prevent the introduction of contraband (firearms, explosives, incendiary devices) intothe protected area as measured against 10 CFR 73.55(d) and the Physical Security Planand Procedures.

12

Site access control activities were observed, including personnel and packageprocessing through the search equipment during peak ingress periods on June 4 and 5,2002. On May 22, 2002, testing of all access control equipment including metaldetectors, explosive material detectors, and X-ray examination equipment was observed. The Access Control event log, an audit, and three self-assessments were also reviewed.

The inspectors also reviewed the Condition Reports (CRs) generated and entered intothe licensee’s corrective action program, to address concerns identified during theinspection. The CRs reviewed are identified in the list of documents contained inAttachment 1 of the report.

b. Findings


4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification

The inspector reviewed the licensee’s performance indicator (PI) data collecting andreporting process as described in procedure SAO-114, “Preparation of NRC and WANOPerformance Indicators.” The purpose of the review was to determine whether themethods for reporting PI data were consistent with the guidance contained in NuclearEnergy Institute (NEI) 99-02, Revision 1 and Revision 2, “Regulatory AssessmentPerformance Indicator Guidelines.” The inspection included a review of the indicatordefinitions, data reporting elements, calculation methods, definition of terms, andclarifying notes for the performance indicators. Plant records and data were sampledand compared to the reported data. The inspector reviewed the licensee’s actions toaddress discrepancies in the performance indicator measurements to verify problemswere satisfactorily resolved.

.1 Safety System Unavailability - Auxiliary Feedwater

a. Inspection Scope (71151)

The inspector reviewed maintenance rule tracking, control room logs, and conditionreports associated with the auxiliary feedwater system. The inspector reviewed plantdata from the 2nd quarter of 2001 through the 1st quarter of 2002 for all three trains ofauxiliary feedwater. The inspector also reviewed two specific condition reports(200110095 and 200110097) related to errors in the data provided to the Nuclear EnergyInstitute prior to issuance to the NRC. The licensee’s quality assurance group hadidentified the errors in the data submittal in response to previous NRC inspectionobservations of data accuracy.

b. Findings


13

.2 Safety System Unavailability - Emergency AC Power


In the 1st Quarter Performance Indicator data, the licensee reported that an engineeringassessment was ongoing to evaluate the 23 emergency diesel generator (EDG) governorfailure modes and effect on fault exposure hours, if any. The inspector reviewed thecondition report (CR No. 200203079) and assessment associated with this failure toevaluate the licensee’s determination that the EDG was fully functional prior to therepairs to correct the governor failure.

b. Findings


.3 Fitness-for-Duty, Personnel Screening, and Protected Area Security Equipment

a. Inspection Scope

The inspectors reviewed the licensee’s programs for gathering and submitting data forthe Fitness-for-Duty, Personnel Screening, and Protected Area Security EquipmentPerformance Indicators. The review included the licensee’s tracking and trendingreports, personnel interviews and security event reports for the Performance Indicatordata collected from the 2nd quarter of 2001 through the 1st quarter of 2002.

b. Findings


4OA3 Identification and Resolution of Problems


The inspector conducted a problem identification and resolution (PI&R) sampleinspection to review Entergy’s actions to address problems related to a portion of theProtected Area boundary that may not have had compensatory measures implementedin accordance with the Security Plan, as documented in CR No. 200112339. Theinspector verified the appropriateness of compensatory measures and actions taken tocount the compensatory hours toward the applicable performance indicator.

b. Findings


4OA4 Inspection Item Follow-up

.1 (Closed) URI 05000247/2001-04-05: Adequacy of auxiliary feedwater (AFW) alignmentand function in response to a feedwater line break. The system response to thepostulated event (feed line break) was beyond the design description documented in the

14

Updated Final Safety Analysis Report. The licensee documented this item in CR No.200106409 and evaluated the auxiliary feedwater system response to a postulated linebreak through an engineering analysis (PSA-010830-01, “Feedwater System Pipe BreakAnalysis). The engineering analysis concluded that decay heat could be removed withthe minimum AFW alignment to ensure that the reactor coolant system does not overpressurize or the pressurizer does not go into a water-solid condition. The inspectorreviewed the analysis and concluded that there was no violation of NRC requirements. This unresolved item is closed.

.2 (Updated) FIN 05000247/01-013-01: Proposed finding due to crew high failure rateduring the 2001 annual requalification simulator examinations. This finding wasdocumented in an October 2001 inspection and initially characterized as a potentialYellow finding, the final safety significance to be determined (TBD). This finding wassubsequently evaluated under the significance determination process (SDP) andcharacterized as (reference NRC to Entergy letters dated December 5, 2001, andFebruary 28, 2002). The 95002 Supplemental Inspection (reference Inspection ReportNo. 50-247/02-09, dated May 31, 2002), assessed the licensee’s evaluation of the crewhigh failure rates and the corrective actions taken to address this performance issue. Asstated in the cover letter to Inspection Report No. 50-247/02-09, this finding remainsopen until after the completion of Entergy’s licensed operator requalificationexaminations, scheduled for September-October 2002, and further review by the NRC. This item remains open.

4OA6 Meetings, Including Exit

On May 24, 2002, the inspectors presented the inspection results of report details 3PP1,3PP2, and 4OA3 to licensee representatives. At that time, the purpose and scope of theinspection were reviewed, and the preliminary findings were presented. The licenseeacknowledged the preliminary inspection findings.

On July 3, 2002, the inspector presented an overall summary of the inspection results toMr. Paul Rubin, and other members of the licensee staff, who acknowledged the findings. No material examined during the inspection should be considered proprietary.

15

ATTACHMENT 1

a. Key Points of Contact

R. Allen Manager, Regulatory AffairP. Asendorf Manager of SecurityT. Barry Security SuperintendentJ. Cambigianis System EngineerF. Dacimo Vice President, OperationsR. Depatie System EngineerW. Durr Assistant Operations ManagerT. Foley System EngineerW. James Maintenance and Construction ManagerJ. McCann Manager, Nuclear Safety and LicensingG. Norton Control Room SupervisorP.K. Parker Maintenance ManagerJ. Perrotta QA ManagerP. Rubin Operations ManagerV. Sacco System Engineer G. Schwartz Director of EngineeringP. Speedling Fire Protection SpecialistR. Taylor QA EngineerM. Vasely System Engineering Section ManagerJ. Ventosa System Engineering Manager

b. List of Items Opened, Closed, and Discussed

Closed

URI 50-247/01-04-05 Auxiliary Feedwater Alignment and Function

Discussed

FIN 50-247/01-013-01 Crew High Failure Rate During 2001 Annual RequalificationSimulator Examinations

c. List of Documents Reviewed

Procedures

COL 21.3, Steam Generator Water Level and Auxiliary Boiler Feedwater SystemCOL 3.1, Chemical and Volume Control SystemSAO-700, Fire Protection and Prevention PolicySAO-703, Fire Protection Impairment Criteria and SurveillanceAOI 28.0.6, Nuclear Side (Outside Containment) FloodingAOI 24.1, Service Water MalfunctionAOI 4.1.1, Loss of Component Cooling WaterARP SJF Window 4-6, Service Water Header High/Low Pressure

Attachment 1 (cont’d) 16

LARP-4, Primary Auxiliary building Sump Pump High LevelEOP ES 1.4, Transfer to Hot Leg RecirculationOASL 15.43, Operator Burden ProgramSAO-206, Temporary Facility ChangeSAL-114, Preparation of NRC and WANO Performance IndicatorsEmergency Plan, Section 5, Rev 01-02aIP-1002, Emergency Notification and Communication, Rev. 25IP-1008, Personnel Radiological Check and Decontamination, Rev. 7IP-1010, Central Control Room, Rev. 3, 4, 5IP1011, Joint News Center, Rev. 6, 7IP-1021, Manual Update, Readout & Printout of PROTEUS Plan Parameter Data, Rev. 6IP-1023, Operations Support Center, Rev. 17, 18IP-1024, Emergency Classification, Rev. 9, 10IP-1027, Personnel Accountability and Evacuation, Rev. 15, 16IP-1047, Obtaining Offsite Exposure Rates From MIDAS Using a Data Terminal, Rev. 8IP-1050, Security, Rev. 2, 3EPMP-EPP-01, Maintenance of Emergency Preparedness, Rev. 14EPMP-EPP-02, Emergency Equipment Inventories and Checklists, Rev. 25EPMP-EPP-04, Emergency Exercise/Drill Procedure, Rev. 8EPMP-EPP-06, Emergency Response Organization Notification Maintenance and Surveillance,Rev. 10

Calculations

PGI-00433, Combustible Loading CalculationPGI-00354-02, Emergency Diesel Generator Heat Exchanger PerformancePSA-010830-01, Feedwater System Pipe Break Analysis

Drawings

9321-F-2736-114, Flow Diagram Chemical and Volume Control System, Sheet 1A208168-51, Flow Diagram Chemical and Volume Control System, Sheet 2

Miscellaneous

Indian Point Nuclear Generating Station Unit 2 Maintenance Rule Basis Document, 13.8 KVSystem, Revision 1Indian Point Nuclear Generating Station Unit 2 Maintenance Rule Basis Document, AuxiliaryFeedwater, Revision 1Indian Point Nuclear Generating Station Unit 2 Maintenance Rule Basis Document, ServiceWater, Revision 1Maintenance Rule Unavailability for 13.8 KV System (4/2000 - 3/2002)Maintenance Rule Reliability for 13.8 KV System (2001 - 2002)Maintenance Rule Unavailability for 440 V System (4/2000 - 3/2002)Maintenance Rule Reliability for 440 V System (2001 - 2002)4th Quarter 2001 System Health Report for 13.8 KV and 440 V SystemsFMX-97-12648-M, Installation of Canopy Seal Clamp AssembliesTFC 2001-100, Gas Turbine 1 Black start Diesel Battery InstallationTFC 2001-033, FCV-1207 Hot Tap Isolation Valve


Condition Reports

200206616, Fire Brigade Leader without SCBA donned200206187, Blocked fire hydrant outside emergency diesel generator building199808862, Debris on 23 diesel generator jacket and lube oil heat exchangers200110318, Epoxy broken off on 21 emergency diesel generator jacket water heat exchanger200203601, 21 and 22 CCW heat exchanger as left performance data200205647, Staff team “X” failed to complete high intensity training200111489, Failure of automatic function of auxiliary feedwater flow control valve FCV-406A200202589, Failure of SWN-7 to fully close due to operator gear stripping200205115, Loss of 13.8 kv alternate safe shutdown bus section 3.0200205369, Failure of safety injection MOV 851B to close200203319, Stroke time for Post accident containment vent valve E-3200205369, Loose wire in MOV 851B open circuit200205673, Large amount of debris on the floor of the 480 volt switchgear room200205433, Incomplete Operability for MOV 851B200109397, Nuclear detector rate comparator indication200109706, Letdown pressure regulator manual operation200110544, Residual heat removal flow control valve position indicator200101809, 22 main feedwater pump steam stop valve control switch200202395, 24 Reactor coolant pump motor temperature indicationCondition Reports (cont.)200110095, Safety system unavailability for auxiliary feedwater error for 3rd quarter data200110097, Safety system unavailability for auxiliary feedwater does not have data available200203079, 23 emergency diesel generator test200112339, Protected area not properly compensated

Sections 3PP1 and 3PP2

Entergy Fitness-for-Duty ProgramFitness-for-Duty Performance Data, January - June 2001; June - December 2001Self-Assessment - Access Control, March 2002, Report 02-001QA Surveillance Report - Fitness-for-Duty Program, April 26, 2002Condition Detail Report 200205250, Individual Failed to Report Immediately for a FFD testCondition Detail Report 200200272, Non-Notification of Employee No Longer Requiring PhotoBadgeCondition Detail Report 200201585, Access not Notified of Employee on Medical Leave; BadgeNot DeactivatedIndian Point Nuclear Generating Station Unit 2 Maintenance Rule Basis Document, 13.8 KVSystem, Revision 1


d. List of Acronyms

AFW auxiliary feedwaterAOI abnormal operating instructionARP annunciator response procedureATWS anticipated transient without scramCFR Code of Federal RegulationsCOL check off listCR condition reportCRDM control rod drive mechanismDBT design basis threatEDG emergency diesel generatorEOP emergency operating procedureFCV flow control valve FFD fitness-for-dutyIPEEE individual plant examination for external eventskv kilovoltL&P lighting and powerLARP local annunciator response procedureNEI Nuclear Energy InstituteNRC Nuclear Regulatory CommissionOASL Operations Administrative SteplistPARS publicly available recordsPI performance indicatorPI&R problem identification and resolutionPM preventive maintenancePWT post-work testRSPS Risk Significant Planning StandardSAO station administrative orderTBD to be determinedTFC temporary facility change